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Title: MATERIALS TESTING REACTOR-ENGINEERING TEST REACTOR TECHNICAL BRANCHES QUARTERLY REPORT, APRIL 1-JUNE 30, 1960

Abstract

@ ? E : : 7 : 8 @ ? 9 8 ; experiment installed was indicated with changes in loading and rod withdrawal sequence developed in the ETRC. The variation in other experimental fluxes in the ETRC due to withdrawal and insention of the two GEANP experiments proposed during operation of the ETR are relatively small. Detailed honizortal neutron flux maps within the 400-g ETR fuel elements establishsd more accurate constants for ex trapolating ETRC fluxes to "full power" ETR values and for determining heat transfer limitations. Comparison of a black and a gray absorber section on control rod No. 13 when partially insented as a regulating rod shows at most only 7% more flux depression for the black section than for the gray. It is shown that, for 400-g ETR fuel elements, 10% more borcu in polyethylene tapes is required in order to be equdvalent to boron uniformly distributed in the coolant space. Changing the metal-to-water ratio of the fuel elements in the ETR from 0.644 to 1.210 without a change in charge life is found to cauae a 20% increase in thermal neutron flux in the inpile experiments for the 11% increase in U/sup 235/ required. Themore » vertical thermal neutron flux distribution in each ETR fuel element was determined during Cycle 27 for the clean core and for the depleted core. Calculations were made of the integrated power following a junior scram in the ETR for varying rod worths to determine the protection afforded by their use. Comparisons of calculated and measured thermal neutron fluxes in the ETRC were made for variation in calculation techniques and reactor physics constants. The full program of 14 capsules containing oxide fuel in the fuel element development program for the Experimental Gas Cooled Reactor is now installed in the ETR with the total burnup now ranging from 500 to 2500 Mwd/MT. Fundamental studies of the metal-water reaction of aluminum--23.4 wt.% uranium ailoy indicate very low reaction rates op to 2300 deg F. Calculations made to maximize the production of U/sup 233/ + Pa/sup 233/ from thorium slugs without exceeding a given heat generation rate indicate that the most efficient method is to use two different fluxes. A preliminary measurement of tbe Co/sup 58/thermal cross section indicates that the 1500 b value makes a sizeable correction in the calculation of fast neutron fluxes from the threshold reaction Ni/sup 58/(n,p)Co/sup 58/. Preliminary total cross section data on Pa/sup 231/ and Pu/sup 241/ were taken with the MTR chopper. Cr ystal spectrometer measurements on the variation of eta for U/sup 233/ in the region 0.01 to 1 ev were compared with eta values obtained from MTR fission and total cross section data and with measurements made in other laboratories. Time-of-flight analyses of Bragg beans from beryllium crystal planes demonstrate the necessity of making such studies prefatory to high precision measurements. An inveatigation made into the system design of the MTR crystal spectrometer shielding cart drive to determine the system response to increasing the drive speed by a factor of 10 indicates that with appropriate design changes a stable system to meet the requirements can be obtained. In the MTR nuclear chemistry program, results from triplicate analyses of the gas produced in highly irradiated beryllium are found to agree with yields of these gases calculated from cross sections and the neutron irradiation history. More accurate values for the half life of Cs/sup 134//sup m/ (2.90 plus or minus 0.01 hr) and for its formation cross section (3.45 plus or minus 0.2 b) were obtained at the MTR. The activation yield ratios of metastable and ground states for Rh/sup 104/ and Ir/sup 192/ were determined at ther mal and at low energy resonances. Results giving comparisons among Au, Mn, and Na thermal and resonance flux monitors show good agreement. The activation thermal cross sections and« less

Publication Date:
Research Org.:
Phillips Petroleum Co. Atomic Energy Div., Idaho Falls, Idaho
OSTI Identifier:
4041744
Report Number(s):
IDO-16648
NSA Number:
NSA-15-015280
DOE Contract Number:  
AT(10-1)-205
Resource Type:
Technical Report
Resource Relation:
Other Information: Orig. Receipt Date: 31-DEC-61
Country of Publication:
United States
Language:
English
Subject:
REACTOR TECHNOLOGY; ABSORPTION; BORON; CONTROL ROD WORTH; DISTRIBUTION; ETR; ETRC; FUEL ELEMENTS; HEAT TRANSFER; MATERIALS TESTING; MTR; NEUTRON FLUX; NEUTRONS; POLYETHYLENES; REACTIVITY; REACTOR FUELING; REACTORS; THERMAL NEUTRONS

Citation Formats

None. MATERIALS TESTING REACTOR-ENGINEERING TEST REACTOR TECHNICAL BRANCHES QUARTERLY REPORT, APRIL 1-JUNE 30, 1960. United States: N. p., 1961. Web. doi:10.2172/4041744.
None. MATERIALS TESTING REACTOR-ENGINEERING TEST REACTOR TECHNICAL BRANCHES QUARTERLY REPORT, APRIL 1-JUNE 30, 1960. United States. doi:10.2172/4041744.
None. Wed . "MATERIALS TESTING REACTOR-ENGINEERING TEST REACTOR TECHNICAL BRANCHES QUARTERLY REPORT, APRIL 1-JUNE 30, 1960". United States. doi:10.2172/4041744. https://www.osti.gov/servlets/purl/4041744.
@article{osti_4041744,
title = {MATERIALS TESTING REACTOR-ENGINEERING TEST REACTOR TECHNICAL BRANCHES QUARTERLY REPORT, APRIL 1-JUNE 30, 1960},
author = {None},
abstractNote = {@ ? E : : 7 : 8 @ ? 9 8 ; experiment installed was indicated with changes in loading and rod withdrawal sequence developed in the ETRC. The variation in other experimental fluxes in the ETRC due to withdrawal and insention of the two GEANP experiments proposed during operation of the ETR are relatively small. Detailed honizortal neutron flux maps within the 400-g ETR fuel elements establishsd more accurate constants for ex trapolating ETRC fluxes to "full power" ETR values and for determining heat transfer limitations. Comparison of a black and a gray absorber section on control rod No. 13 when partially insented as a regulating rod shows at most only 7% more flux depression for the black section than for the gray. It is shown that, for 400-g ETR fuel elements, 10% more borcu in polyethylene tapes is required in order to be equdvalent to boron uniformly distributed in the coolant space. Changing the metal-to-water ratio of the fuel elements in the ETR from 0.644 to 1.210 without a change in charge life is found to cauae a 20% increase in thermal neutron flux in the inpile experiments for the 11% increase in U/sup 235/ required. The vertical thermal neutron flux distribution in each ETR fuel element was determined during Cycle 27 for the clean core and for the depleted core. Calculations were made of the integrated power following a junior scram in the ETR for varying rod worths to determine the protection afforded by their use. Comparisons of calculated and measured thermal neutron fluxes in the ETRC were made for variation in calculation techniques and reactor physics constants. The full program of 14 capsules containing oxide fuel in the fuel element development program for the Experimental Gas Cooled Reactor is now installed in the ETR with the total burnup now ranging from 500 to 2500 Mwd/MT. Fundamental studies of the metal-water reaction of aluminum--23.4 wt.% uranium ailoy indicate very low reaction rates op to 2300 deg F. Calculations made to maximize the production of U/sup 233/ + Pa/sup 233/ from thorium slugs without exceeding a given heat generation rate indicate that the most efficient method is to use two different fluxes. A preliminary measurement of tbe Co/sup 58/thermal cross section indicates that the 1500 b value makes a sizeable correction in the calculation of fast neutron fluxes from the threshold reaction Ni/sup 58/(n,p)Co/sup 58/. Preliminary total cross section data on Pa/sup 231/ and Pu/sup 241/ were taken with the MTR chopper. Cr ystal spectrometer measurements on the variation of eta for U/sup 233/ in the region 0.01 to 1 ev were compared with eta values obtained from MTR fission and total cross section data and with measurements made in other laboratories. Time-of-flight analyses of Bragg beans from beryllium crystal planes demonstrate the necessity of making such studies prefatory to high precision measurements. An inveatigation made into the system design of the MTR crystal spectrometer shielding cart drive to determine the system response to increasing the drive speed by a factor of 10 indicates that with appropriate design changes a stable system to meet the requirements can be obtained. In the MTR nuclear chemistry program, results from triplicate analyses of the gas produced in highly irradiated beryllium are found to agree with yields of these gases calculated from cross sections and the neutron irradiation history. More accurate values for the half life of Cs/sup 134//sup m/ (2.90 plus or minus 0.01 hr) and for its formation cross section (3.45 plus or minus 0.2 b) were obtained at the MTR. The activation yield ratios of metastable and ground states for Rh/sup 104/ and Ir/sup 192/ were determined at ther mal and at low energy resonances. Results giving comparisons among Au, Mn, and Na thermal and resonance flux monitors show good agreement. The activation thermal cross sections and},
doi = {10.2172/4041744},
journal = {},
number = ,
volume = ,
place = {United States},
year = {1961},
month = {1}
}